ML17219A274

Proposed Tech Specs,Reflecting Deletion of Core Barrel Movement.Safety Evaluation & Determination of NSHC Encl

ML17219A274
Person / Time
Site:	Saint Lucie
Issue date:	12/18/1986
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17219A273	List: L-86-515, Application for Amend to License DPR-67,revising Tech Specs to Reflect Deletion of Core Barrel Movement.Fee Paid ML17219A274
References
NUDOCS 8612240195
Download: ML17219A274 (13)

Text

INOEX LIMITING CONOITIOX FOR OPERATION AND SURVEILLANCE RE UIREMENTS 5ECTNN 3/4.4.4 RESSURI2ER......,...,,......................

P 3/4.4 5

STBW GENERATORS................................,...,

PAGE 3/4 4-4 3/4 4-5 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE...... "" " "

. 3/4 4-12 3/4.4.7 3/4.4.8 Leakage Oetectfon Systems.......

Reactor Coolant System Leakage..

CHEMISTRY.......................

SPECIFIC ACTIVITY......."......

3/4 4-12 3/4 4-14 3/4 4-15

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4~1 7 3/4.4.9 PRESSURE/TEMPERATURE LIMITS........................ -... 3/4 4-21 Reactor Coolant System..........

Pressurizer......................

3/4 4-21 3/4 4-25 3/4.4.10 STRUCTURAL INTEGRITY........~.........................

3/4 4-26 3/4-4 11 Safety Class 1 Components.......

Safety Class 2 Components.......

Safety Class 3 Components.......

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3/4 4-26 3/4 4-37 3/4 4-53

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~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4~56 3/4.4.12 PORV BLOCK VALVES......................................

3/4 4-58 3/4.4.13 POMER OPERATED RELIEF VALVES...........................

3/4 4-59 3/4.4.14 REAC QR COOLANT PUMP - STARTING........................

3/4 4-60 3/4.4.15 REACTOR COOLANT SYSTEM VENTS...............,......-;.....

3/4 4-61 3 4.5 EMERGENCY CORE COOLING SYSTEMS ECCS 3/4 5-1 3/4 5-3 3/4 5.2 ECCS SUBSYSTEMS Tavg

> 325'F.....

3/4.5.3 ECCS SUBSYSTEMS - Tang

<< 325'F.....

3/4.5.4 REFUELING MATER TANK.............. ~

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PDR ST. LUCIE-UNIT 1 V

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Mo. gg, yN, m 3/4.5.1 SAFETY INJECTION TANKS..........'......

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REACTOR COOLANT SYSTEM y~(~~~8 r

3.4.11 r

barrel movement all be limited o less than the plitude

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o bali 0

tribution (APQ) nd Spectral Ana sis (SA)

A t Leve t

app ic le THERMAL POMER level.

PLIC I

TY ACT h:A'i e A.

d/or SA excee ng thei applicable, Alert Levels, POW WT 0 may, proce provid the following'actions are taken

~ r AP sh ll ba asured d processed at least on per 4

u SA sh b

me d wlthln 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> and at least once ho s thee er'an SA shall be processed at least o

per da$s, A S cia Re rt idea 'fying the cause(s) exceeding placable 1 t Lgv I ~ shalg be prepare'nd sub-mitted to the C

i sioh ursuant to S~ifica on

.9.2 w

hi 30 spoof detection. ~

Mi+ the,APQ an r SA ewe~Hing their applicable Ac ion Lev@s, withi ho~s reduce.7HER+JL PC'E'ER BY 25".o of RATED THE."

PQME.

and denopshate through monitorina o

the co neutr de ters, that, APO and SAi a

be o

ced o below e r appli bleiAlert Le

.limits in T

S OB w

hi the nex 6 M rs Mi t

mea u

levels A 0 and/

A fering fro their bas in lese by re tha 1

Special port descr.

ing he m as

,ed le els hall be e

red an&,su itted to the

, C ioniqursu nt t5 Sp sf at>

n 6.9.2'wit

$p 10 days ov p

essxng

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ST.

LUCIE - UNIT 1

3/4 4 56

Basell o

teria CoreWrre movement Alert be s and Ac on Lave

~s as detematned by AP0 and SA n$ tor$ ng of Alfe encore neut n de ctors, shall b

determined at nom al THE POWE@levels of 20%

0", 80" and 100 TEQ THERMAL POWEib,du ng the rene'ter sta p test program; t se Al t Levels and Ac Levels shall b

pe'ported Iri'a specla aport pur ant to specsflcat 6.9.2 wlthln 31 days after ini ally reaching 1

" of 0 THECAL OWER.

4.4.11.2 tine 'Noni tor in Core barrel movement shall b

determined be 1

than the R

and SA Al t LeveD by using the exc e neutron de rs to measure A

and S

at the foll ing frequencies:

a.

b.

APO.data shall e measured and process at least once p

7 days S

data all be measu d and processed at east once at nomi 1

THE POWER levels of 0, 50~,

80~ and 10 of RATEO THER'NL PO ter each refueling d at least once p

4 months ereaft 4.4

.3 e orts The suIts of all perko Lc APD and SA mon$ t Ing shall included tne Annual crating Report fob the period fn whic the monitoring was erformed.

ST.

LUCIE - UNIT 1 3/4 4-57

REACTOR COOLANT SYSTEM."I 8ASKS The nondestructive 'testing for repairs on components greater than 2

inches diameter gives a high degree of confidence,in the integrity of the system, and will detect any significant defects.in and near the new welds.

Repairs on components 2 inches in diameter or smaller receive a.

surface examination which assures a similar standard of integrity.

In each case, the leak test will ensure leak tightness, during normal.

operation.

For normal opening and reclosing, the structural integrity of the Reactor Coolant System is unchanged.

Therefore, satisfactory pe, or-...ance of a system leak test at 2235 psia following each ooening and subsequent reclosing is acceptable demonstration of the system's s.ructural inta-rity.

These leak tests will be conducted within the pressure-temperature limitations for Inservice Leak and Hydrostatic Testing and Figure 3.4-2.

The Safety Class 2 and 3 components will be pressure tested at least once toward the end of each inspection interval (10 years).

The Safety Class 2 ccmponents having a design temperature above 4CO'.F will be pressure tested at not less than 125 percent of the system design pressure while those components having a design temperature of 400'F and, belo~ will be pressure.

tested at 110 percent of design pres:ure.

The Safety Class 3 components will be pressure tested at the levels indicat~d in Specification 4.4.10.3b.

3/4.4.11 De le%.J his specifica is provid ensure early te

'on o

ex essi core b

rel mover t

1 ii snould ur.

Core barrel element will '"

e.e t ~.by u

g four excore

~tron tectors i

. ta',n";-..pI',"iu P

babilii Qistribution f.

and Speci Ana sis (SA).

Bas ne or ba el movement Al~t.'els and Acti evels at nomi THi3'QL POll evels or 205, 5, 80" nd 1005 RAi 0 THERttAL P ER wz 1

be erma ed during e reactor s

rtu est pro' mod catio to the. re red onitoring og m may be justifie y

a ys of th dat tained a

by a examina on of the affected art 'ring t. e plant

.utdown at the e,

or the first uel cycle.

T. LUCIE - UNIT 1

8 3/4 4-13

ATTACHMENT 2 SAFETY EVALUATION Introduction During licensing of St. Lucie Unit I, a problem was identified at Palisades and several other Combustion Engineering

reactors, including St.

Lucie Unit I,

concerning the core barrel hold-down ring design.

NRC addressed this problem in Section 3.9.I of the St. Lucie Unit I Safety Evaluation Report (SER) dated November 8, l974, and stated in the SER, "A monitoring program will be required until either a modification has been made to the internals or data indicates the program may be discontinued."

The St. Lucie Unit I core barrel hold-down ring was redesigned to provide additional force to hold the core barrel in place, and in Supplement I to the SER, NRC stated that the redesigned ring was acceptable and that the issue was resolved with incorporation of a surveillance program to monitor core barrel movement.

This NRC position is reflected in the Bases for the Core Barrel Movement Technical Specification in the statement, "A

modification to the required monitoring program may be justified by an analysis of the data obtained and by an examination of the affected parts during the plant shutdown at the end of the first fuel cycle.

Discussion The St. Lucie Unit I Technical Specification for Core Barrel Movement was required by NRC as described above.

The purpose of the Technical Specifications was to verify the effectiveness of the redesigned core barrel hold-down ring by determining the core barrel movement baseline and by monitoring core barrel movement against the baseline.

Reporting requirements were also included as part of the core barrel movement Technical Specifications.

By letter dated April 22, l977 (L-77-I22), FPL submitted to NRC the Core Barrel Movement Baseline Report as required by Technical Specification 4.4.I I.I. The baseline was established by monitoring core support barrel motion at nominal power levels of 20, 50, 80, and I00 percent of rated thermal power during the reactor startup test program.

As stated in the report, the core support barrel is moving less than + 8.8 mils amplitude motion (99.7 percent confidence level) at the snubber gap level.

The baseline monitoring results provided sufficient verification of the effectiveness of the redesigned core barrel hold-down ring, in that Palisades had experienced approximately 300 mils amplitude motion as determined from the measured wear of the snubber blocks.

However, St. Lucie Unit I has continued monitoring core barrel movement in accordance with the Technical Specifications.

Results of the monitoring program have been included with the Annual Operating Reports, beginning with the l977 report.

Based on a review of the results presented in these nine reports, it can be seen that the core barrel motion has been as expected.

Furthermore, upon identification of the thermal shield problem in Spring l983, the core barrel was removed, inspected and, where

damaged, repaired.

During the post repair inspection, all six snubber blocks were examined and there were no indications of excessive core barrel movement.

RJS2/002/I

IR eN I

IN

Conclusions The redesigned core barrel hold-down ring has eliminated the possibility of excessive core barrel movement such as that which occurred at Palisades.

This has been verified by more than nine years of core barrel movement monitoring and by physical inspection of the core barrel snubber blocks.

FPL believes that the purpose of the Technical Specification has been satisfied and that an adequate basis has been provided to justify deletion of the Technical Specification without endangering the health and safety of the public.

RJS2/002/2

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ATTACHMENT3 DETERMINATION OF NO SIGNIFICANTHAZARDS CONSIDERATIONS The standards used to arrive at a determination that a request for amendment involves no significant hazards consideration are included in the Commission's regulations, 10 CFR 50.92, which states that no significant hazards considerations are involved if the operation of the facility in accordance with the proposed amendment would not (I) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possiblity of a new or different kind of accident from any accident previously evaluated or (3) involve a significant reduction in a margin of safety.

Each standard is discussed as follows:

(I)

Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.

The purpose of the Core Barrel Movement Technical Specification was to verify the effectiveness of the redesigned core barrel hold-down ring by determining the core barrel movement baseline and by monitoring core barrel movement against the baseline.

The baseline was determined, and monitoring core barrel movement has been performed for over nine years of plant operation.

The results have shown that excessive core barrel movement is not possible with the redesigned core barrel hold-down ring.

Because core barrel movement monitoring has been shown to be no longer necessary, and because core barrel movement is not considered in the accident analyses, operation of St. Lucie Unit I without a requirement for core barrel movement monitoring will not involve an increase in the probability or consequences of an accident previously evaluated.

(2)

Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed amendment would only delete the requirement for core barrel movement monitoring, and would not alter any of the assumptions or methodologies used in the safety analyses.

Furthermore, there is no change to the operation of the plant so that a new or different kind of accident is not possible as a result of this change.

(3)

Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

The Core Barrel Movement Technical Specification does not establish any margins of

safety, and therefore, deletion of the requirement for monitoring core barrel movement will not result in a reduction in a margin of safety.

RJS2/002/3

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Based on the above, we have determined that the amendment request does not (I) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the probability of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.

Furthermore, the proposed amendment is similar to Example (iv) of arnendrnents that are considered not likely to involve significant hazards considerations, identified in the staff procedure for determination of no significant hazards, in that the proposed arnendm'ent would constitute relief granted upon demonstration of acceptable operation from an operating restriction that was imposed because acceptable operation was not yet demonstrated.

Therefore, the proposed amendment involves no significant hazards consideration.

RJS2/002/4

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